1. Field Of The Invention
The present invention relates to liquid moderated thermal reactors and, more particularly, to a method for operating a light water nuclear reactor resulting in improved fuel utilization and neutron economy and reduced control requirements.
2. The Prior Art
General methods of improving fuel utilization in nuclear reactors have been suggested and evaluated but are not immediately applicable to light water reactors of current design, or are impractical from a technological or economic viewpoint. One of the more promising reactor concepts is the spectral shift controlled reactor (SSCR), described in U.S. Pat. No. 3,081,246, which issued Mar. 12, 1963, to M. C. Edlund. This reactor operates with a mixture of heavy and light water as both coolant and moderator. Improves fuel utilization and a portion of the reactor control requirement are achieved by controlling the relative concentrations of light and heavy water in the coolant, and thereby changing the neutron slowing down power or moderation to change the average neutron energy in the reactor, and when fertile material is present to increase the productive resonance capture of neutrons by the fertile material. At the start of a fuel cycle, there is a high concentration of D.sub.2 O in the core and the neutron spectrum is shifted toward higher energies, which reduces reactor excess reactivity and increases the reactor conversion ratio. As operation proceeds, H.sub.2 O is added and the spectrum is shifted to lower energies, which increases reactivity and tends to reduce the conversion ratio. Unlike the present invention, the SSCR is based on continuous shifts in the neutron spectrum (characterized by the H/fuel ratio) throughout the lifetime of the fuel in the reactor. The present invention does not rely upon D.sub.2 O to shift the neutron spectrum and does not require large changes in H/fuel ratio during each cycle of operation. As will be discussed in detail below, the neutron spectrum is shifted from an exceptionally low value used for initial operation to an exceptionally high value by withdrawing fuel rods from the reactor core and thereby reducing the relative amount of fuel in the core.
There are a number of disadvantages associated with the SSCR, most of which derive from the use of D.sub.2 O, which are avoided in the present invention. The D.sub.2 O and the equipment required to handle it are expensive and a new supply of D.sub.2 O is needed for each fuel cycle. The use of D.sub.2 O results in the production of large quantities of tritium which are difficult to control and represent a potential radiological hazard. Since the coolant is changed from high concentration D.sub.2 O to H.sub.2 O each fuel cycle, the higher conversion ratio characteristic of large D.sub.2 O concentrations is only achieved during a portion of the fuel cycle (i.e., the average coolant D.sub.2 O concentration during the cycle is about one-half the initial and final values). At the start of operation of each cycle when the coolant is high concentration D.sub.2 O, the neutron leakage is larger than in a 100% water moderated reactor due to the smaller slowing down power of D.sub.2 O as compared to H.sub.2 O. Neutron leakage from the core is reflected as a direct loss in conversion ratio, thus reducing the improvement in conversion ratio which could otherwise have been obtained by a shift in neutron energy spectrum.
Other patents to M. C. Edlund in this area are: U.S. Pat. No. 3,142,624, wherein the same technique described above is used in conjunction with a seed-blanket breeder reactor; and U.S. Pat. No. 3,247,072 wherein control is achieved by varying the density of a hydrogen-isotope-bearing vapor in the core.
Outside of shifting the neutron spectrum, the art takes various approaches to improving fuel utilization. Many of these approaches involve designing the reactor core with different spatial regions of different reactive systems. In U.S. Pat. No. 3,884,886 to Crowther, for example, plutonium fuel is placed in fuel assemblies at the periphery of the core where there is a low energy neutron spectrum, and enriched fuel assemblies are placed at the center portions of the core where there is a higher energy neutron spectrum. U.S. Pat. No. 3,141,827 to Ishenderian discloses a breeder reactor core with a central core of enriched fuel with alternate zones of depleted and enriched fuel. This arrangement causes a peak thermal flux to occur in the depleted elements which results in increased conversion of U-238 to PU-239. U.S. Pat. No. 3,093,563 to Menke teaches a reactor core constructed with an inner active core in the fast neutron range and an outer active core in the slow range. U.S. Pat. No. 3,351,532 to Raab, Jr., et al. teaches a seed-blanket reactor wherein the H/fuel ratio in the seed is set so the number of fissions in the seed is a maximum and the H/fuel ratio in the blanket is set so the number of conversions in the blanket is maximized. Unlike any of the latter group of reactors, the present invention does not rely upon a reactor core which is constructed with different spatial regions of different reactivity. Again, none of these inventions teaches the art of setting the H/fuel ratio at a low value for an initial period of operation and later increasing the H/fuel ratio by withdrawing rods and thereby reducing the relative amount of fuel in the core.
In addition to the foregoing, the prior art recognizes that the reactivity of a fuel assembly can be increased by withdrawing fuel rods, and this operation has been considered at the time the fuel assembly is normally scheduled for discharge in order to boost its reactivity and allow for some additional period of operation and thereby obtain some improvement in fuel utilization. This prior art does not allow for assembly operation in an undermoderated condition and, consequently, does not acquire the benefits of increased fissile material production and reduced control requirements accorded the present invention, both of which result in a much larger improvement in fuel utilization than can be obtained operating in a near optimum moderated condition throughout the life of the fuel assembly and withdrawing fuel rods at the time normally scheduled for discharge.